Cold Water Injection and Rod Ejection Analysis of Annular Fueled PWRs by a Hybrid Lumped Parameter Model

2014 ◽  
Author(s):  
Juliana P. Duarte ◽  
José de Jesus Rivero ◽  
Antonio Carlos M. Alvim ◽  
José Roberto C. Piqueira ◽  
Paulo F. F. Frutuoso e Melo
Keyword(s):  
Solid Fuel ◽  
Cold Water ◽  
Hot Spot ◽  
Water Injection ◽  
Lumped Parameter Model ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Steady State Condition ◽  
Lumped Parameter

Annular fuels are being studied to increase the power of advanced third-generation reactors by 50%. This paper aimed to analyze transient scenarios through a hybrid lumped parameter-finite difference model in a pressurized water reactor with annular fuel. The model used in this work is more detailed than the double lumped parameter one, but still simple enough to model some transients in PWR fuels, as rod ejection accident and cold water insertion accident. The heat transfer equations are solved by the numerical semi-implicit Crank-Nicolson method together with point kinetics equations with six groups of delayed neutrons and a lumped parameter model for the reactor coolant. The model takes into account in an approximate way the hot spot by using a composed peaking factor equal to 2.5. The reactivity feedback is taken into account by considering the Doppler effect of fuel temperature, and also moderator temperature variation. The results were compared with solid fuel performance and showed that the annular fuel reached considerable lower fuel temperature profiles even for 150% power, as compared to 100% power for solid fuel, thus showing that this kind of fuel has a better safety performance for the transients analyzed. The rod ejection accident showed that feedback effects can lead the reactor to a new safe steady state condition.

scholarly journals Analysis of thermal hydraulic behavior of KONVOI PWR during a design extension condition

2021 ◽  
pp. 4-4
Author(s):  
Salwa Helmy ◽  
Magy Kandil ◽  
Ahmed Refaey
Keyword(s):  
Power Plants ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
Hydraulic Behaviour ◽  
Loss Of Coolant ◽  
Safety Systems

In Nuclear Power Plants the Design Extension Conditions are more complex and severe than those postulated as Design Basis Accidents, therefore, they must be taken into account in the safety analyses. In this study, many hypothetical investigated transients are applied on KONVOI pressurized water reactor during a 6-in. (182 cm2) cold leg Small Break Loss-of-Coolant-Accident to revise the effects of all safety systems ways through their availability/ nonavailability on the thermal hydraulic behaviour of the reactor. The investigated transients are represented through three cases of Small Break Loss-of-Coolant-Accident as, case-1, without scram and all of the safety systems are failure, case-2, the normal scram actuation with failure of all safety systems (nonavailability), and finally case 3, with normal actuation scram sequence and normal sequential actuation of all safety systems (availability). These three investigated transient cases are simulated by creation a model using Analysis of Thermal-Hydraulics of LEaks and Transient code. In all transient cases, all types of reactivity feedbacks, boron, moderator density, moderator temperature and fuel temperature are considered. The steady-state results are nearly in agreement with the plant parameters available in previous literatures. The results show the importance effects of the feedbacks reactivity at Loss-of-Coolant-Accident on the fallouts power, since they are considered the key parameters for controlling the clad and fuel temperatures to maintain them below their melting point. Moreover, the calculated results in all cases show that the thermal hydraulic parameters are in acceptable ranges and encounter the safety criterion during Loss-of-Coolant-Accident the Design Extension Conditions accidents processes. Furthermore, the results show that the core uncovers and fuel heat up do not occur in KONVOI pressurized water reactor in theses the Design Extension Conditions simulations since, all safety systems provide adequate core cooling by sufficient water inventory into the core to cover it.

scholarly journals Visualization and Finding Ways to Intensify the Processes of Non-isothermal Mixing Flows

2018 ◽  
Vol 3 (3) ◽  
pp. 240
Author(s):  
Sataev A.A. ◽  
Duntsev A.V.
Keyword(s):  
Fluid Dynamics ◽  
Cold Water ◽  
Density Structure ◽  
Pressurized Water Reactor ◽  
Thermal Reliability ◽  
Reactor Plant ◽  
Pressurized Water ◽  
Real Geometry ◽  
Computational Fluid Dynamics Cfd ◽  
Isothermal Mixing

Simulation of mixing flows of different temperature, density structure has important implications for the assessment of thermal reliability of reactor plants, thermo-cyclic pulsations, and safety analysis. To study the mixing model was used for the mixing, which was visualized by using imaging methods. The injection of cold water into the hot volume was examined, which simulates the flow of the coolant in the pressurized-water reactor. The obtained results have given the basis for further analysis of non-isothermal mixing flows. However, the model is still far from the real geometry of the reactor plant. The construction of a reactor reduced model with a simulation of one loop of a coolant flow with low settings has been developed for a more detailed study of the processes of non-isothermal mixing flows has planned. In the future, these data will be used in the programs of computational fluid dynamics (CFD). 

A Case Study of a Valve Failure Due to Condensation-Induced Waterhammer

2002 ◽  
Author(s):  
Henry E. Harling
Keyword(s):  
Cold Water ◽  
Lower Pressure ◽  
Bubble Collapse ◽  
Horizontal Line ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Initial Flow ◽  
Flash Tank ◽  
Pressure Spike ◽  
Moisture Separator

In preparation for a power reduction at a Pressurized Water Reactor (PWR) power plant, the Moisture Separator Drain Tanks (MSDT) were being transitioned from pumping to a flash tank to dumping to the Condenser. This evolution includes initially transitioning to pressure feeding to a lower pressure flash tank first. As soon as the MSDT Pump was tripped, a waterhammer occurred in the 6″ line that contained a valve that isolated the flow path to the lower pressure flash tank. The waterhammer caused the operator of this valve to fail due to a fracture of the yoke on the actuator. Several thermal-hydraulic mechanisms were evaluated and it was postulated that the failure mechanism was a bubble collapse in a 6″ horizontal line upstream of the failed isolation valve leading to the lower pressure (a.k.a., “D”) flash tank. A prerequisite for this mechanism is that the isolation valve leading to the “D” flash tank is “tight” such that line is “cold.” As the valve opens the initial flow regime would be annular before transitioning to a dispersed regime. The vapor and cold water interacted leading to the collapse of an assumed 1″ diameter “trapped” bubble. The collapse of the postulated bubble yielded an unbalanced impulse force of 6000 lbs. for approximately 5 – 10 milliseconds. The pressure spike that generated the unbalanced force was estimated at 208 psid. The initial acceleration of the pipe was estimated to have been as high as 21 g, which was greater than the 11 g estimated necessary to fail the valve actuator.

scholarly journals Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor

2011 ◽  
Vol 32 (4) ◽  
pp. 67-79
Author(s):  
Tomasz Bury
Keyword(s):  
Nuclear Power ◽  
Large Scale ◽  
Nuclear Reactor ◽  
Computer Code ◽  
Hydrogen Combustion ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Lumped Parameter ◽  
Loss Of Coolant

Thermodynamic consequences of hydrogen combustion within a containment of pressurized water reactor Gaseous hydrogen may be generated in a nuclear reactor system as an effect of the core overheating. This creates a risk of its uncontrolled combustion which may have a destructive consequences, as it could be observed during the Fukushima nuclear power plant accident. Favorable conditions for hydrogen production occur during heavy loss-of-coolant accidents. The author used an own computer code, called HEPCAL, of the lumped parameter type to realize a set of simulations of a large scale loss-of-coolant accidents scenarios within containment of second generation pressurized water reactor. Some simulations resulted in high pressure peaks, seemed to be irrational. A more detailed analysis and comparison with Three Mile Island and Fukushima accidents consequences allowed for withdrawing interesting conclusions.

scholarly journals Neutronic analysis of fuel pin design for the long-life core in a pressurized water reactor

2021 ◽  
Vol 11 (1) ◽  
pp. 9-15
Author(s):  
Van Khanh Hoang ◽  
Vinh Thanh Tran ◽  
Dinh Hung Cao ◽  
Viet Ha Pham Nhu
Keyword(s):  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Water Reactor ◽  
Fuel Model ◽  
Reactivity Coefficients ◽  
Fuel Pin ◽  
Safety Parameters ◽  
Long Life

This work presents the neutronic analysis of fuel design for a long-life core in a pressurized water reactor (PWR). In order to achieve a high burnup, a high enrichment U-235 is traditionally considered without special constraints against proliferation. To counter the excess reactivity, Erbium was selected as a burnable poison due to its good depletion performance. Calculations based on a standard fuel model were carried out for the PWR type core using SRAC code system. A parametric study was performed to quantify the neutronically achievable burnup at a number of enrichment levels and for a numerous geometries covering a wide design space of lattice pitch. The fuel temperature and coolant temperature reactivity coefficients as well as the small and large void reactivity coefficients are also investigated. It was found that it is possible to achieve sufficient criticality up to 100 GWd/tHM burnup without compromising the safety parameters.

A Consistent, Hybrid-Dynamical-System, Lumped-Parameter Model of Tire–Terrain Interactions

2015 ◽  
Vol 10 (3) ◽  
Author(s):  
Pravesh Sanghvi ◽  
Harry Dankowicz
Keyword(s):  
Dynamical System ◽  
Shear Failure ◽  
Internal State ◽  
Numerical Study ◽  
Lumped Parameter Model ◽  
State Variables ◽  
Hybrid Dynamical System ◽  
Steady State Condition ◽  
Internal State Variables ◽  
Lumped Parameter

This paper establishes the internal mathematical and energetic consistency of a hybrid-dynamical-system, lumped-parameter, planar, physical model for capturing transient interactions between an elastically deformable tire and an elastically deformable terrain as a baseline result for more realistic models that account for permanent deformation, shear failure, and three-dimensional contact conditions. The model accounts for radial and circumferential deformation of the tire as well as normal and tangential deformation of the terrain. It captures the onset and loss of contact as well as localized stick and slip phases for each of the discrete tire elements by a suitable evolution of a collection of associated internal state variables. The analysis characterizes generic transitions between distinct phases of contact uniquely in forward time and proves that all internal state variables remain bounded during compact intervals of contact. The behavior of the model is further illustrated through an analytical and numerical study of two instances of tire-terrain interactions under steady state condition.

scholarly journals Core Design of a Small Pressurized Water Reactor with AP1000 Fuel Assembly Using SRAC and COBRA-EN Codes

2020 ◽  
Vol 2020 ◽  
pp. 1-7
Author(s):  
Van Khanh Hoang
Keyword(s):  
Safety Margin ◽  
Coolant Temperature ◽  
Pressurized Water Reactor ◽  
Fuel Temperature ◽  
Pressurized Water ◽  
Control Rod ◽  
The Core ◽  
Small Modular Reactor ◽  
Active Core ◽  
And Performance

This paper presents the core design and performance characteristics of a 300 MWt small modular reactor (SMR) with fuel assemblies of the AP1000 reactor. Numerical calculations have been performed to evaluate a proper active core size and core loading pattern using the SRAC code system with the JENDL-4.0 data library and the CORBRA-EN code. The calculated temperature coefficients including fuel temperature, coolant temperature, and isothermal temperature coefficient provide adequate negative reactivity feedbacks. The thermal-hydraulic analysis reveals acceptable radial and axial fuel element temperature profiles with significant safety margin of fuel and clad surface temperature. A safety analysis using the CORBRA-EN code shows that the core will remain covered during the entire transient procedure of the fast transient of remarkably increasing power that would be caused by the ejection of control rod. The analysis results indicate that the core with a cycle length of 2.22 years is achievable while satisfying the operation and safety-related design criteria with sufficient margins.

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